ML20236F112
| ML20236F112 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 06/25/1998 |
| From: | Eric Simpson Public Service Enterprise Group |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20236F114 | List: |
| References | |
| LCR-H98-03, LCR-H98-3, LR-N98277, NUDOCS 9807020029 | |
| Download: ML20236F112 (17) | |
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e 3
l Pubic Servk:e Dactro and Gas
- compary E. C. simpson Pubhc Service Electric and Gas Company P O. Box 236. Hancocks Bndge. NJ 08038 609-339-1700 Semot Va Present - Nxits Enyommg l JUN 2 51998 LR-N98277 L
LCR H98-03 i United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Gentlemen:
REQUEST FOR CHANGE TO TECHNICAL SPECIFICATIONS DELETION OF AUTOMATIC DEPRESSURIZATION SYSTEM VALVE STARTUP TESTING L HOPE CREEK GENERATING STATION FACILITY OPERATING LICENSE NPF-57 DOCKET NO. 50-354 L in accordance with 10CFR50.90, Public Service Electric & Gas (PSE&G) Company j hereby requests a revision to the Technical Specifications (TS) for the Hope Creek l Generating Station (HC). In accordance with 10CFR50.91(b)(1), a copy of this submittal has been sent to the State of New Jersey.
I implementation of the prc posed changes contained in this submittal will delete the l - requirement to perform !n-situ functional testing of the Automatic Depressurization 1 System (ADS) valves once every 18 months as part of startup testing activities. The proposed changes in this submittal are similar to a TS amendment approved by the NRC in an SER dated January 29,1997, for PECO Energy's Limerick Generating l Station. NRC review of the changes contained in this submittalis requested by
/ j l
L December 15,1998 to support the next refueling outage (RFO8) at Hope Creek. /
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The proposed changes have been evaluated in accordance with 10CFR50.91(a)(1), AO()/
using the criteria in 10CFR50.92(c), and a determination has been made that this request involves no significant hazards considerations. The basis for the requested change is provided in Attachment 1 to this letter. A 10CFR50.92 evaluation, with a determination of no significant hazards consideration, is provided in Attachment 2. The marked up Technical Specification pages affected by the proposed changes are provided in Attachment 3. ;j 9807020029 9 W 5 yDR ADOCK 05000354i
= n h NvNa" raper
Document Control Desk JUN 2 51998
,LR-N98277 Upon NRC approval of this proposed change, PSE&G requests that the amendment be made effective on the date of issuance, but allow an implementation period of sixty days to provide sufficient time for associated administrative activities.
Should you have any questions regarding this request, we will be pleased to discuss them with you.
Sincerely, ,
Affidavit Attachments (3) l 95-4933
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Document Control Desk -
LR-N98277 M 2 51998 C Mr. H. Miller, Administrator - Region i U. S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Mr. R. Ennis -
Licensing Project Manager - Hope Creek U. S. Nuclear Regulatory Commission One White Flint North Mail Stop 14E21 11555 Rockville Pike Rockville, MD 20852 Mr. S. Pindale (X24)
USNRC Senior Resident inspector- HC Mr. K. Tosch, Manager IV '
Bureau of Nuclear Engineering P. O. Box 415 Trenton, NJ 08625 I
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- a REF: LR-N98277
- LCR H98-03 STATE OF NEW JERSEY )
SS.
COUNTY OF SALEM E. C. Simpson, being duly sworn according to law deposes and says:
I am Senior Vice President - Nuclear Engineering of Public Service Electric and Gas Company, and as such, I find the matters set forth in the above referenced letter, concerning Hope Creek Generating Station, Unit 1, are true to the best of my knowledge, information and belief.
~MM (/
Subscribed and n to before me thi d N d .At L ,1998 iem huA< W ftary Public of dewhrsey My Commission expires on / / ,[ / tr / 2 $
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L______ ____--_-________ _
l D:cument Control D3ck LR-N98277 Attcchment 1 LCR H98-03 HOPE CREEK GENERATING STATION FACILITY OPERATING LICENSE NPF-57 DOCKET NO. 50-354 REVISIONS TO THE TECHNICAL SPECIFICATIONS (TS)
BASIS FOR REQUESTED CHANGE:
Public Service Electric and Gas Company (PSE&G), under Facility Operating License No. NPF-57 for the Hope Creek Generating Station requests that the TS contained in Appendix A to the Operating License be amended as proposed herein to revise TS Surveillance Requirement 4.5.1.d.2.b to delete the requirement to perform in-situ functional testing of the Automatic Depressurization System (ADS) valves once every 18 months as part of startup testing activities. The proposed changes to the TS surveillance requirements are indicated on the marked-up TS pages contained in Attachment 3 of this submittal. The NRC approved a similar change in an SER dated January 29, 1997, for PECO Energy's Limerick Generating Station.
REQUESTED CHANGE, PURPOSE AND BACKGROUND:
Currently, Hope Creek TS Surveillance Requirement 4.5.1.d.2.b requires that at least once every 18 months the ADS safety relief valves (SRVs) undergo manual in-situ functional exercise testing '
as part of startup testing activities following an outage. This .
testing is typically performed at 750 +/- 50 psig. This proposed -
TS change will revise the wording in TS Surveillance Requirement l 4.5.1.d.2.b to delete the requirement to perform the in-situ '
testing, but will provide a requirement to verify that when the valves are tested pursuant to the requirements of TS Section 4.0.5, which pertains to inservice inspection and testing of .
American Society of Mechanical Engineers (ASME) Code Class 1, 2, !
and 3 components, that each ADS valve is capable of being opened.
Performing these tests imposes an unnecessary challenge on these valves and has been linked to valve degradation of the associated pilot valves. The proposed TS change will ensure that the valves ,
will continue to be tested pursuant to the requirements specified )
in TS 4.0.5, and that each ADS valve is capable of being opened. ;
Operability of the ADS valves is assured through the remaining !
- surveillance tests and inspections that are routinely performed L on these valves. This proposed TS change should reduce leakage l through these valves, thereby improving their reliability by reducing the potential for spurious valve actuation.
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Drcument Centrol Dack LR-N98277 Attachment 1 LCR H98-03
'In addition, the changes to TS Surveillance Requirement
.4.5.1.d.2.b.also require a change to the existing TS Surveillance Requirement.4.4.2.1. Specifically, the-18 month channel calibration for the SRV acoustic monitor will no longer require an exception to the provisions of TS 4.0.4,'nor adjustments to full open noise levels. As described in the following section, Hope. Creek will perform SRV acoustic monitor channel calibration in a manner that no longer requires reactor steam pressure or noise level adjustments. Therefore, the notes associated with TS Surveillance Requirement 4.4.2.1 are no longer needed.
JUSTIFICATION OF REQUESTED CHANGES:
' System Description The ADS utilizes five of 14 main steam line SRVs to reduce !
reactor _ pressure vessel pressure during small pipe breaks, f or after containment isolation, in the event that the High j Pressure Coolant Injection (HPCI) system and/or Reactor j Core Isolation Cooling (RCIC) system fail to maintain adequate reactor pressure vessel water level. The ADS is independent of any other Emergency Core Cooling System (ECCS), and is designed to reduce reactor vessel pressure in order to permit the low pressure ECCS (i.e., Low i Pressure Coolant Injection (LPCI) and/or Core Spray (CS) I systems) to inject water into the reactor vessel in time to cool the core and limit fuel cladding temperature.
The ADS utilizes the following five main steam line SRVs:
1SNPSV-FO13A ISNPSV-F013B 1SNPSV-FOl3C 1SNPSV-F013D 1SNPSV-F013E The five ADS SRVs are controlled by'ECCS logic to automatically open and provide rapid reactor pressure vessel depressurization in order to enable low pressure ECCS cooling. Reactor operators may initiate ADS manually as required by plant Emergency Operating Procedures (EOPs).
These valves, and the remaining nine non-ADS SRVs, are two stage pilot actuated relief valves manufactured by the
. Target Rock Corporation. The nine non-ADS SRVs may also be actuated manually to relieve reactor pressure vessel Page 2 of 9
l Document Centrol Dack LR-N98277 Attachment 1 LCR H98-03
' pressure. All 14 SRVs are designed to be opened by either of the following two methods:
- 1) Automatically - Steam pressure above the setpoint pressure overcomes the spring forces on the pilot disc assembly and opens the pilot disc; or
- 2) Manually - Direct actuation by a pneumatic diaphragm l assembly (i.e., auxiliary-actuation device) that overcomes the pilot. spring force allowing the pilot' disc to open.
The SRVs are designed such that once the pilot valve is opened, a force imbalance across the main valve disc piston is created causing the main valve disc to rapidly and reliably open at greater than 50 psig nominal inlet pressure.
i All 14 SRVs are included in the Hope Creek Inservice Testing (IST) Program,-which was developed to ensure q compliance with Section IX of the ASME Boiler and Pressure 1 Vessel (B&PV) Code. These SRVS are tested in accordance )
with ASME OM-1, 1987, which provides guidance for testing relief / safety valves,.and Hope Creek TS Section 3/4.4.2,
" Safety / Relief Valves.". As stated in the NRC SER for Hope Creek TS Amendment No. 64, dated January 27, 1994, this 4 testing requires that 50% of the SRV pilot stage assemblies be set: pressure tested in accordance with the manufacturer's recommendations every 18 months and that all 14 SRV pilot stage assemblies are tested at least once per 40 months. The SRVs are removed, set pressure tested and ;
reinstalled or replaced with spares that have been set 1
-pressure' tested and stored in accordance with the l manufacturer's recommendations. This testing is performed j off-site at a certified test facility (e.g., Wyle Labs or l NWS) and is supported, as needed, by the valve manufacturer !
-(Target Rock Corporation). However, based on the history I of SRV setpoint performance, all 14 SRVs have been replaced during the refueling outages at Hope Creek. This replacement schedule is not expected to. change until setpoint performance improves.
PSE&G also performs additional SRV surveillance testing, which together fulfills the periodic testing requirements stipulated'in ASME OM-1, 1987, Section 3.3. The five ADS Page 3 of 9 l
D cument'Cantrol Desk LR-N98277 Att chment 1 LCR H98-03
'SRVs'are maintained and tested by PSE&G as ASME " Class 1 Pressure Relief Valves with Auxiliary Actuating Devices" as stipulated in ASME OM-1, 1987, Sections 3.3.1.1 and 3.4.1.1.
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In addition to'the required surveillance tests and ;
inspections described above, Hope Creek routinely inspects the-SRVs (as recommended by General Electric Co. Service Information Letter (SIL) No. 196, " Summary of Recommendations for Target Rock Main Steam Safety / Relief Valves"), refurbishes, and setpoint tests the entire SRV assembly at the certified test facility. This SRV inspection and testing is currently performed as part of routine SRV preventative maintenance and control activities as specified by TS Surveillance Requirement 4.4.2.3.
Pilot valve assemblies replaced during the refueling outages are as-found tested, repaired (as necessary) and as-left tested. The testing performed by the qualified j test: facility is in accordance with their procedures, which are reviewed by PSE&G. The procedures institute Target Rock Corporation testing requirements. PSE&G specifies the setpoint and leakage-test pressure. Target Rock Corporation, in accordance with their procedures, performs any repairs. Repairs include cleaning, lubrication, machining, lapping, and installation of new gaskets and parts as required'. The qualified test facility and Target Rock Corporation provide test and repair reports for the work performed. Target Rock Corporation works under PSE&G's ASME Code,Section XI, Repair and Replacement Program.
Main valve bodies are disassembled and inspected after l
' decontamination. Minimum maintenance includes renewal of gasket, lapping of seat and disc, and leak testing with nitrogen. Parts occasionally needing replacement are the valve disc, seat, spring, piston rings and fasteners.
PSE&G implements the appropriate inspection guidance specified in General Electric SIL No. 196. Target Rock Corporation provides repair and parts reports to PSE&G, and any repairs are performed in accordance with Target Rock Corporation's procedures, which are reviewed by PSE&G.
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After any repairs, main valves are steam tested for stroke andupost-test' leakage in accordance with the qualified test Page 4 of 9 c l L !
Document Control Desk LR-N98277 Attachment 1 LCR H98-03
'faciiity's test procedure. The qualified test facility provides a test report, which includes test results. Main valve testing is a restricted flow test. The valve outlet is restricted so that only limited steam flow is permitted.
The valve disc travel is not restricted by this test, but steam flow is limited by a gag device. Test stand supply and steam recovery capability requires limited flow testing.
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Valve Performance History ADS is considered to be one or the most highly reliable Boiling Water Reactor (BWR) ECCS systems. As stated in PECO Energy's March 26, 1996, submittal (approved by the NRC in the aforementioned January 29, 1997, SER), there has never been a failure of an ADS valve to open once the pilot i-- valve has opened during in-situ functional testing at any BWR. In addition, PECO Energy had stated that SRV main disc failures are considered extremely rare and statistically insignificant, and that no such failures had occurred with an in-service SRV. PSE&G has confirmed that to date, there have been no additional failures of an ADS valve to open once the pilot valve has opened during in-situ functional testing, nor have there been main disc failures with an in-se c-ice SRV. General Electric and Target Rock had indicated in SIL 196, Supplement 17,
" Target Rock SRV Main Disc Spring Relaxation and Tip Breakage," that valve dynamics under full flow conditions (i.e., discharge not gagged) are much less severe that those under limited flow conditions. Under full flow conditions, flow forces and discharge backpressure slow the opening of the main valve disc and is thought to contribute less significantly to spring relaxation, which could result in valve failure.
Hope Creek has experienced instances of SRV leakage and has instituted leakage monitoring action plans through measurement of SRV' tail pipe temperatures. Tail pipe temperature " Alert" and " Action" levels were developed to ensure that adequate time is available to schedule an outage and replace the affected SRV before reaching a temperature at which an SRV has been known to lift due to severe pilot valve erosion. Therefore, PSE&G considers that all unnecessary challenget to the SRV, which could aggravate the leakage problem, be absolutely minimized if Page 5 of 9
D2cument Centrol Dock LR-N98277 Attcchment 1 LCR H98-03
'adeqbate proof of operability can be provided by alternate means. This approach is consistent with a BWR Owners Group (BWROG) study of NUREG-0737, " Clarification of TMI Action Plan Requirements," Item II.K.3.16, " Reduction of Challenges and Failures of Relief Valves," which recommends that in-situ ADS valve actuations be reduced as much as possible.
The BWROG, of which PSE&G is an active participant, has a committee which is tasked with understanding the maintenance and surveillance practices that affect SRV leakage. The results from this effort should ultimately lead to changes in these practices and also in SRV design to mitigate the SRV leakage problems, which are not confined to ADS valves.
The table below' documents recent instances of SRV leakage at Hope Creek following plant startups, of which a high percentage occurred after performance of the in-situ ADS /SRV functional-test (i.e., TS Surveillance Requirement 4.5.1.d.2.b).
Cycle 5 Data (5/94-8/95)
Position "E" S/N 357 leaking from startup Position "G" S/N 358 leaking from startup Position "K" S/N 365 leaking after plant scram 5/15/94 Cycle 6 Data: (3/96-9/97)
Position "D" S/N 349 leaking from startup Position "R" S/N 364 leaking from'startup Position "F" S/N 354 Started leaking 11/96 stopped 3/6/97 Position."L" S/N 345 Started leaking 11/96 Cycle 7 Data: (12/97 present)
Position "A" S/N 352 leaking from startup Position "G" S/N 343 leaking from startup Position "E" S/N 348 leaking from startup l;
Adequacy of Existing and Alternate Testing i PSE&G believes that the in-situ testing of the ADS /SRVs is not necessary because the remaining ADS surveillance tests i
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Dacument Centrol D3ck LR-N98277 Attcchment 1 LCR E98-03
- and SRV inspections provide the necessary assurance of ADS valve operability. These additional tests and inspections of the ADS /SRVs are described below.
1 i ADS Logic System Functional Test This test, performed during each refueling cycle, verifies the ECCS. logic functions to actuate the ADS on Low Reactor Water Level - Level 1, and High Drywell Pressure. Verification of ADS from the start of the automatic initiation logic to, but not including, instrument gas / accumulator solenoids is demonstrated.
It is important to note that the TS Bases for this functional test do not require actual stroking of the ADS /SRV.
Steam Relief Valve Cyclic Testinc This test, performed each refueling outage, verifies proper operation of the ADS solenoid valves, air operator, and pilot assembly.
ADS Leak Test This test, performed during each refueling cycle and each time maintenance is performed on the ADS valve, verifies that ADS instrument gas / accumulator leakage is low enough to ensure that there will be sufficir.c
. pneumatic pressure for design basis ADS /SRV operation.
The ADS design basis calls for two ADS /SRV actuations at 70% of the maximum drywell pressure (62 psig) to depressurize the reactor pressure vessel down to-the Residual Heat Removal (RHR) Shutdown Cooling operating pressure' range.
l l SRV Setpoint/ Leakage Testing i l
l- These functional tests and inspections are performed on at least 50% of the SRV pilot stage assemblies-during each refueling outage. These tests verify the i pilot valve and setpoint spring assembly open and l close at the required set-pressure, and that leakage i is within strict vendor specified criteria.
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Dgcument Centrol Deck LR-N98277 Attachment 1- LCR H98-03
' Main Disc Exercise Test
-SRV main. disc actuation and leakage is also verified when the entire valve assembly is shipped to the certified test facility. In addition, the testing verifies at least once every five years that all of the SRV main' discs can freely open as specified in TS Surveillance Requirement 4.4.2.3.
Unlike the approved PECO Energy submittal, Hope Creek's proposed elimination of the in-situ test during startup as described'in this TS change request will still retain the current method of stroke time testing the ADS valves as described in an NRC SER dated January 2, 1992, for the Hope Creek Generating Station. Since stroke time testing is currently performed as part of the offsite pilot stage assembly tests each refueling cycle, the proposed changes will have no impact on this test. Unlike the PECO TS changes, this current practice of measuring stroke' times will still be retained as a method of detecting valve degradation. In addition, valve degradation can also be reliably detected by the other alternative testing and inspection methods described above. The combined tests and inspections discussed above, and retention of the ADS /SRV stroke time tests, verify the required ADS critical component performance requirements. The proposed TS change will effectively only reduce the frequency of verifying that the. opening of the pilot disc directly results in opening of the main valve disc. However, this ADS /SRV sub-component function is considered to be extremely reliable based upon the simplicity of this aspect of SRV design.
This is supported by the ADS /SRV valve performance history described above.
In addition, the Bases for the Improved Standard TS (i.e.,
NUREG-1433,. Revision 1, dated April 7, 1995) states that the in-situ ADS testing also verifies that the SRV discharge line is not blocked. The probability of blocking an SRV line and preventing ADS depressurization is considered to be extremely remote. Similar to PECO Energy, PSE&G has an effective Foreign Material Exclusion (FME)
Program in place at Hope Creek in order to minimize the potential of debris blocking an ADS /SRV discharge line.
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Document Control Desk LR-N98277 Attechment 1 LCR 998-03
'ThL elimination of the in-situ testing ADS /SRVs also requires that PSE&G change the manner in which the channel calibration is performed for the SRV acoustic monitors. The channel calibration required by TS Surveillance Requirement 4.4.2.1.b will still be performed on an 18 month frequency; however, the channel
-calibration of acoustic monitors will be accomplished by use of a signal' generator input both by millivolt input and " ping" test of acoustic monitors to verify. channel functionality and response within the necessary range and accuracy. This channel calibration method is similar to that utilized at PECO Energy's Limerick Generating Station and will be implemented at Hope Creek under the provisions of 10CFR50.59. Since the SRV acoustic-monitor channel calibration will be performed in a manner that no 11onger requires reactor steam pressure, SRV opening or adjustments to full open noise level, the notes associated with TS Surveillance Requirement 4.4.2.1, which provide a compliance exception from the provisions of TS 4.0.4 and an allowance for noise level adjustments, are no longer needed and should be removed.
Conclusion PSE&G considers that the in-situ testing imposes an unnecessary challenge on the ADS /SRVs and has been linked I
to SRV degradation (e.g., pilot valve and/or main valve leakage). Therefore, this proposed TS change should reduce SRV leakage:and improve ADS /SRV reliability by reducing.the potential for spurious SRV actuation.
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Occument Control Dask LR-N98277 Attcchment 2 LCR H98-03 HOPE CREEK GENERATING STATION FACILITY OPERATING LICENSE NPF-57 DOCKET NO. 50-354 REVISIONS TO THE TECHNICAL SPECIFICATIONS (TS) 10CFR50.92 EVALUATION Public Service Electric & Gas (PSE&G) has concluded that the proposed changes to the Hope Creek Generating Station (HC)
Technical Specifications.do not involve a significant hazards consideration. In support of this determination, an evaluation of each of the three standards set forth in 10CFR50.92 is provided below.
REQUESTED CHANGE The proposed changes affect TS Surveillance Requirement 4.5.1.d.2.b by deleting the requirement to perform in-situ functional testing of the Automatic Depressurization System (ADS) safety relief valves (SRVs) during startup testing activities, and eliminates a note concerning an exception to TS 4.0.4 associated with TS Surveillance Requirement 4.4.2.1.b.
BASIS
- 1. The proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.
The proposed TS change does not involve any physical changes to plant structures, systems or components (SSC).
The ADS will continue to function as designed. The ADS is an Emergency Core Cooling System (ECCS) designed to mitigate the consequences of an accident, and therefore, can not contribute to the initiation of any accident. The ADS utilizes five of the 14 main steam line SRVs as the primary method for depressurizing the reactor pressure vessel to permit low pressure core cooling capability in the. event of a small break Loss-of-Coolant-Accident (LOCA) if the high pressure cooling systems (i.e., High Pressure Cooling Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) systems) fail to maintain adequate reactor vessel water level.
Deleting the TS surveillance requirements to perform the in-situ testing of the ADS /SRVs during startup, as proposed, should reduce the probability of an inadvertent opening of an SRV as discussed in Section 15.1.4 of the Page 1 of 4
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Document Control Doak LR-N98277 l Attcchment 2 LCR H98-03 )
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' Hope Creek UFSAR since deleting this testing requirement will eliminate a known initiator of SRV pilot leakage and subsequent erosion. This proposed TS change will have a tendency to increase, rather than decrease, the reliability of the ADS /SRVs by eliminating the in-situ ADS functional startup testing. The probability of the ADS /SRVs to open on demand has been demonstrated to be extremely high and is not measurably improved through the in-situ ADS functional l startup testing.
Using the provisions of 10CFR50.59, PSE&G will establish a method for performing SRV acoustic. monitor channel calibration that does not require reactor steam pressure.or SRV opening. This testing method will comply with the current TS definition of CHANNEL CALIBRATION. Since the notes associated with TS Surveillance Requirement 4.4.2.1 (providing a compliance exception to the provisions of TS 4.0.4 to allow for proper reactor steam pressure to perform
.the test and an allowance for noise level adjustments) are no longer needed, their removal will not affect plant operation.or testing and will not involve an increase in the probability or consequences'of an accident previously evaluated.
This proposed TS change will not increase the probability of occurrence of a malfunction of any plant equipment important to safety. Alternate testing methods at Hope Creek and at the offsite test facility adequately demonstrate proper ADS valve operation and assure that the valves will continue to function as designed. Existing.
surveillance testing and inspections of the ADS /SRVs at Hope Creek verify that the ADS initiation logic, solenoid valve operation, pneumatic gas supply integrity and air operator assembly (including pilot rod) will operate as !
designed. Offsite testing verifies pilot disc operation, setpoint calibration, stroke time and main valve disc i operation.
Deleting the in-situ testing requirement, as proposed, will reduce the probability of increasing SRV leakage, which i .should reduce the probability of an inadvertent opening of an SRV. .Therefore, any SRV pilot leakage that can be
, eliminated would reduce the probability of occurrence of a malfunction of that SRV. Deleting the ADS /SRV in-situ functional test will in no way increase any consequences of a malfunction of plant equipment important to safety. The consequences of a malfunction of an ADS /SRV.as discussed in-the Hope Creek UFSAR remain unchanged.
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l Dscument centrol Dock LR-N98277 Attachment 2 LCR H98-03 l.
'In addition, eliminating a known initiator of SRV leakage, as proposed in this TS change, would help reduce operator workarounds.in the form of suppression pool cooling and letdown operation activities. As a result, this will reduce the unnecessary operation of the Residual Heat Removal (RHR) and its supporting systems.
~Therefore, the proposed TS change does not involve an increase in the probability or consequences of an accident j previously evaluated.
- 2. The proposed change does not create the possibility of a new i or different kind of accident from any accident previously l evaluated.
The proposed TS changes do not involve any physical changes to plant SSC. The design and operation of the ADS /SRVs are not changed from that currently described in the UFSAR.
The ADS will continue to function as designed to mitigate the consequences of an accident. No' changes of any kind l are being made to the valves, auxiliary components or ADS logic. Deleting the requirement to perform the ADS in-situ functional test during plant startup as proposed in this TS change request reduces the likelihood of an SRV developing a leak and degrading throughout the. subsequent operating cycle. Therefore, there is no possibility that
-implementing this proposed TS change would create a different type of malfunction to the ADS /SRVs than any previously evaluated.
Eliminating the requirement to perform the in-situ testing of the ADS /SRVs during startup activities does not create a new or different-type of accident than any previously
. evaluated. There is no accident-scenario associated with testing the ADS /SRVs other than.the inadvertent opening of a relief valve, which is currently discussed in Section 15.1.4 of the UFSAR. The' proposed TS changes do not alter the conclusions described in the UFSAR regarding an inadvertent opening of an SRV. No new or different type of
-accident will be created as a result of these proposed changes.
Therefore, the proposed TS change does not create the possibility of a new or different kind of accident from any previously evaluated.
Using the provisions of 10CFR50.59, PSE&G will establish a method for performing SRV acoustic monitor. channel l
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Dncument Centrol D20k LR-N98277 Attachment 2 LCR H98-03
' calibration that does not require reactor steam pressure or SRV opening. This testing method will comply with the current TS definition of CHANNEL CALIBRATION. Since the notes associated with TS Surveillance Requirement 4.4.2.1 (providing a compliance exception to the provisions of TS 4.0.4 to allow for proper reactor steam pressure to perform the test and an allowance to perform noise level adjustments) are no longer needed, their removal will not affect plant operation or testing and will not create the possibility of a new or different kind of accident from any previously evaluated.
- 3. The proposed change does not involve a significant reduction in a margin of safety.
The proposed TS change involves deleting the requirement to perform in-situ functional testing of the ADS /SRVs during startup activities. This testing imposes an unnecessary challenge on the ADS /SRVs and has been linked to SRV degradation (e.g., pilot valve and/or main valve leakage).
This proposed TS change should reduce SRV leakage and improve ADS /SRV reliability by reducing the potential for spurious SRV actuation. Since ADS operability can be readily demonstrated with extremely high confidence by the existing surveillance tests and inspections performed for the ADS, there will be no reduction in any margin of safety resulting from this proposed TS change. Therefore, the proposed TS change does not involve a significant reduction in a margin of safety.
Using the provisions of 10CFR50.59, PSE&G will establish a method for performing SRV acoustic monitor channel calibration that does not require reactor steam pressure or SRV opening. This testing method will comply with the current TS definition of CHANNEL CALIBRATION. Since the notes associated with TS Surveillance Requirement 4.4.2.1 (providing a compliance exception to the provisions of TS 4.0.4 to allow for proper reactor steam pressure to perform the test and an allowance to perform noise level adjustments) are no longer needed, their removal will not affect plant operation or testing and will not involve a significant reduction in a margin of safety.
CONCLUSION Based on the above, PSE&G has determined that the proposed changes do not involve a significant hazards consideration.
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